• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.207-219
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• 2022

The suppression of high-intensity blob structures in the scrape-off layer (SOL) by ion-cyclotron range of frequencies (ICRF) power, leading to a decrease in the turbulent fluctuation level, is observed first in the Experimental Advanced Superconducting Tokamak (EAST) experiment. This suppression effect from ICRF power injection is global in the whole SOL at EAST, i.e. blob structures both in the regions that are magnetically connected to the active ICRF launcher and in the regions that are not connected to the active ICRF launcher could be suppressed by ICRF power. However, more ICRF power is required to reach the full blob structure suppression effect in the regions that are magnetically unconnected to the active launcher than in the regions that are magnetically connected to the active launcher. Studies show that a possible reason for the blob suppression could be the enhanced E_r × B shear flow in the SOL, which is supported by the shaper radial gradient in the floating potential profiles sensed by the divertor probe arrays with increasing ICRF power. The local RF wave power unabsorbed by the core plasma is responsible for the modification of potential profiles in the SOL regions.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2311-2320
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• 2022

The main outcomes of the experiments H2P6 performed in the thermal-hydraulics large-scale PANDA facility at PSI in the frame of the OECD/NEA HYMERES-2 project are presented in this article. The experiments of the H2P6 series consists of two PANDA tests characterized by the activation of three (H2P6_1) or one (H2P6_2) cooler(s) in an initially stratified and pressurized containment atmosphere. The initial stratification is defined by a helium-rich region located in the upper part of the vessel and a steam/air atmosphere in the lower part. The activation of the cooler(s) results i) in the condensation of the steam in the vicinity of the cooler(s), ii) the corresponding activation of large scale natural circulation currents in the vessel atmosphere, with the result of iii) the re-distribution and mixing of the Helium stratification initially located in the upper half of the vessel and iv) the continuous pressure decay. The initial helium layer represents hydrogen generated in a postulated severe accident. The main question to be answered by the experiments is whether or not the interaction of the different, localized cooler units would be important for the application of numerical methods. The paper describes the initial and boundary conditions and the experimental results of the H2P6 series with the suggestion of simple scaling laws for both experiments in terms of i) the temperature difference(s) across the cooler(s), ii) the transient steam and helium content and iii) the pressure decay in the vessel. The outcomes of this scaling indicate that the interaction between separate, closely localized units does not play a prominent role for the present experiments. It is therefore reasonable to model several units as one large component with equivalent heat transfer area and total water flow rate.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3034-3042
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• 2022

The results of radioecological monitoring of the cooling pond Beloyarsk NPP (Russia) have been presented. The influence of waste technological waters of thermal and fast NPP reactors on the content of artificial radionuclides in surface waters and bottom sediments of the Beloyarsk reservoir has been studied. The long-term dynamics of the specific activity of ⁶⁰Co, ⁹⁰Sr, ¹³⁷Cs and ³H in the main components of the freshwater ecosystem at different distances from the source of radionuclide discharge has been estimated. Critical radionuclides (⁶⁰Co and ¹³⁷Cs), routes of their entry and periods of maximum discharge of radioisotopes into the cooling pond have been determined. It is shown that the technology of electricity generation at Beloyarsk NPP, based on fast reactors, has a much smaller effect on the flow of artificial radionuclides into the freshwater ecosystem of the reservoir. During the entire period of monitoring studies, the decrease in the specific activity of radionuclides from NPP origin in surface waters was 4.3-74.5 times, in bottom sediments 10-505 times. The maximum discharge of artificial radionuclides into the reservoir was noted during the period of restoration and decontamination work aimed at eliminating emergencies at the AMB thermal reactors of the first stage of the Beloyarsk NPP.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3631-3640
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• 2022

Uranium mill tailing ponds (UMTPs) are risk source of debris flow and a critical source of environmental U and Rn pollution. The technology of microbial induced calcium carbonate precipitation (MICP) has been extensively studied on reinforcement of UMTs, while little attention has been paid to the effects of MICP on U & Rn release, especially when incorporation of metakaolin and bacillus subtilis (MBS). In this study, the reinforcement and U & Rn immobilization role of MBS -MICP solidification in different grouting cycle for uranium mill tailings (UMTs) was comprehensively investigated. The results showed that under the action of about 166.7 g/L metakaolin and ~50% bacillus subtilis, the solidification cycle of MICP was shortened by 50%, the solidified bodies became brittle, and the axial stress increased by up to 7.9%, and U immobilization rates and Rn exhalation rates decrease by 12.6% and 0.8%, respectively. Therefore, the incorporation of MBS can enhance the triaxial compressive strength and improve the immobilization capacity of U and Rn of the UMTs bodies solidified during MICP, due to the reduction of pore volume and surface area, the formation of more crystals general gypsum and gismondine, as well as the enhancing of coprecipitation and encapsulation capacity.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1676-1685
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• 2021

We develop a new high-fidelity multiphysics model to simulate boron chemistry in the porous Chalk River Unidentified Deposit (CRUD) deposits. Heat transfer, capillary flow, solute transport, and chemical reactions are fully coupled. The evaporation of coolant in the deposits is included in governing equations modified by the volume-averaged assumption of wick boiling. The axial offset anomaly (AOA) of the Seabrook nuclear power plant is simulated. The new model reasonably predicts the distributions of temperature, pressure, velocity, volumetric boiling heat density, and chemical concentrations. In the thicker CRUD regions, 60% of the total heat is removed by evaporative heat transfer, causing boron species accumulation. The new model successfully shows the quantitative effect of coolant evaporation on the local distributions of boron. The total amount of boron in the CRUD layer increases by a factor of 1.21 when an evaporation-driven increase of soluble and precipitated boron concentrations is reflected. In addition, the concentrations of B(OH)₃ and LiBO₂ are estimated according to various conditions such as different CRUD thickness and porosity. At the end of the cycle in the AOA case, the total mass of boron incorporated in CRUD deposits of a reference single fuel rod is estimated to be about 0.5 mg.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3879-3891
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• 2021

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4134-4145
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• 2022

The collimator is one of the high-heat-flux components used to avoid a series of vacuum and thermal problems. In this paper, the heat load distribution throughout the collimator is first calculated through experimental data, and a transient thermodynamic simulation analysis of the original model is carried out. The error of the pipe outlet temperature between the simulated and experimental values is 1.632%, indicating that the simulation result is reliable. Second, the model is optimized to improve the heat transfer performance of the collimator, including the contact mode between the pipe and the flange, the pipe material and the addition of a twisted tape in the pipe. It is concluded that the convective heat transfer coefficient of the optimized model is increased by 15.381% and the maximum wall temperature is reduced by 16.415%; thus, the heat transfer capacity of the optimized model is effectively improved. Third, to adapt the long-pulse steady-state operation of the experimental advanced superconducting Tokamak (EAST) in the future, steady-state simulations of the original and optimized collimators are carried out. The results show that the maximum temperature of the optimized model is reduced by 37.864% compared with that of the original model. The optimized model was changed as little as possible to obtain a better heat exchange structure on the premise of ensuring the consumption of the same mass flow rate of water so that the collimator can adapt to operational environments with higher heat fluxes and long pulses in the future. These research methods also provide a reference for the future design of components under high-energy and long-pulse operational conditions.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.169-179
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• 2023

Filtered containment system is a passive safety system that controls the over-pressurization of containment in case of a design-based accidents by venting high pressure gaseous mixture, consisting of air, steam and radioactive particulate and gases like iodine, via a scrubbing system. An indigenous lab scale facility was developed for research on iodine removal by venturi scrubber by simulating the accidental scenario. A mixture of 0.2 % sodium thiosulphate and 0.5 % sodium hydroxide, was used in scrubbing column. A modified mathematical model was presented for iodine removal in venturi scrubber. Improvement in model was made by addition of important parameters like jet penetration length, bubble rise velocity and gas holdup which were not considered previously. Experiments were performed by varying hydrodynamic parameters like liquid level height and gas flow rates to see their effect on removal efficiency of iodine. Gas holdup was also measured for various liquid level heights and gas flowrates. Removal efficiency increased with increase in liquid level height and gas flowrate up to an optimum point beyond that efficiency was decreased. Experimental results of removal efficiency were compared with the predicted results, and they were found to be in good agreement. Maximum removal efficiency of 99.8% was obtained.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.566-574
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• 2023

Investigating the seepage characteristics of the leaching solution in the ore-bearing layer during the in situ leaching process can be useful for designing the process parameters for the uranium mining well. We prepared leaching solutions of four different viscosities and conducted experiments using a self-developed multifunctional uranium ore seepage test device. The effects of different viscosities of leaching solutions on the seepage characteristics of uranium-bearing sandstones were examined using seepage mechanics, physicochemical seepage theory, and dissolution erosion mechanism. Results indicated that while the seepage characteristics of various viscosities of leaching solutions were the same in rock samples with similar internal pore architectures, there were regular differences between the saturated and the unsaturated stages. In addition, the time required for the specimen to reach saturation varied with the viscosity of the leaching solution. The higher the viscosity of the solution, the slower the seepage flow from the unsaturated stage to the saturated stage. Furthermore, during the saturation stage, the seepage pressure of a leaching solution with a high viscosity was greater than that of a leaching solution with a low viscosity. However, the permeability coefficient of the high viscosity leaching solution was less than that of a low viscosity leaching solution.